Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 40277 | 25 October 2003 04:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degradation of the Reactor Coolant System (Rcs) Pressure Boundary | During inspection of the hot leg nozzles on RCP loop #2 in accordance with the Alloy 600 Program Plan during Refuel 12, evidence of leakage was identified on nozzle RC-IPT-0106B. This constitutes degradation of the RCS pressure boundary. Corrective action is being evaluated. RC-IPT-0106B is a pressure transmitter that taps off of reactor coolant system hot leg (loop) # 2 . The location of the Pressure transmitter is found in Waterford Unit 3, FSAR Chapter 5.1, Flow Diagram Reactor Coolant System, Figure 5.1-3. The license said that the picture indicates a small amount of boric acid crystals in the area of the leakage. Licensee did not know if the leakage is from a weld, pin hole leak, etc. The NRC Resident Inspector was notified of this event by the licensee. | Reactor Coolant System | 05000382/LER-2003-003 |
| ENS 40278 | 27 October 2003 01:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Indication of Leakage | Inspection of the pressurizer bottom nozzles in accordance with the Alloy 600 Program during Refuel 12 identified a wet stain on heater sleeve C1, and a nozzle with a white substance extruding from a portion of the penetration on heater sleeve C3. The substance could not be verified at the time, and further investigation is required to determine if the substance is boron on heater sleeve C3. This condition constitutes a degradation of the RCS Pressure Boundary and is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector will be notified of this event by the licensee. | 05000382/LER-2003-003 | |
| ENS 40295 | 4 November 2003 06:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Number of Steam Generator Tubes Requiring Repair Exceeds 1 Percent | At 0400 CDT on November 4, 2003 with Waterford 3 refueled, steam generator inspection was being performed in accordance with Technical Specification (TS) 4.4.4.2, Steam Generator Tube Sample Selection and Inspection. The results of the on-going inspection performed on Steam Generator 31 indicated that greater than 1% of the inspected tubes require repair. The number of inspected tubes for SG 31 is 8906 and to date 98 tubes require repair. Based on TS 4.4.4.2.C, this determination results in the Category C-3. The current inspection results do not meet the criteria specified for steam generator tube degradation in NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73. However, for Category C-3, TS Table 4.4-2, Steam Generator Tube Inspection, requires that the results of the inspection be reported under 10 CFR 50.72(b)(2), which was superceded by an 8 hr notification in accordance with 50.72(b)(3)(ii). Per TS 4.4.4.5, the final results of the SG tube inspection, which fall into Category C-3, shall be reported in a special report within 30 days and prior to resumption of plant operation. The licensee has notified the NRC Resident Inspector. | Steam Generator | |
| ENS 40306 | 10 November 2003 11:00:00 | 10 CFR 73.71(b)(1), Safeguards Event | Safeguard Vulnerability - Physical Security | Discovered vulnerability in a safeguard system that could allow access to a controlled access area for which compensatory measures were not employed. The licensee notified the NRC Resident Inspector. TAS notified Region IAT member who will followup for NRC. Contact the Headquarters Operations Officer for additional details.
After further review it was determined that there was no vulnerability that would allow access to a controlled access area, therefore the licensee is retracting this event. The NRC Resident Inspector was notified. | ||
| ENS 40519 | 13 February 2004 11:25:00 | Other Unspec Reqmnt | 24 Hour Notification Per Licensee Condition 2.F | On February 13, 2004, Waterford 3 reactor power reached an estimated maximum reactor power level of 101.9% due to Atmospheric Dump Valve #2 (ADV#2) failing full open. ADV#2 failed open due to the control system setpoint failing low. Operations personnel promptly responded to the condition by reducing turbine load (5 megawatts per minute ) and manually closing ADV#2, to restore plant power. ADV#2 was open for less than a minute. Per the FSAR, this transient is an analyzed event and is bounded by the safety analyses. This event is being reported in accordance with Waterford 3 event notification and reporting procedure, UNT-006-010, which is more conservative than the guidance provided in NRC Inspection Procedure 61706. This event is being reported pursuant to Waterford 3 License Condition 2.F as a 24 hour notification. The licensee notified the NRC Resident Inspector.
On February 13, 2004, Waterford 3 provided a 24 hour notification to the NRC (reference EN# 40519) stating that due to a failed open atmospheric dump valve, plant reactor power level momentarily reached a value of 101.9%. The event was reported conservatively based on the guidance provided in the plant's event notification and reporting procedure. The procedure has been revised to incorporate the guidance provided by NRC inspection Procedure 61706, 'Core Thermal Power Evaluation.' Based on this guidance, reporting of events such as stated above (transient power increases) are not required unless reactor power level exceeds 102%. Accordingly, this event is not reportable as power level only reached 101.9% momentarily. Based on the above, the 24 hour notification made on 2/13/2004 is being retracted. The licensee has notified the NRC Resident Inspector. Notified R4 DO (J. Clark) | ||
| ENS 40632 | 1 April 2004 01:19:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | New Worst Case Single Failure May Exceed 10 Cfr 50.46 Acceptance Criterion for Sbloca | A new worst case single failure was identified on 03-31-04 for the small break LOCA analysis of record due to a postulated loss of a DC power bus. This condition may result in the analysis of record exceeding 10 CFR 50.46 acceptance criterion of 2200 F when a small break LOCA with this failure is postulated. The current small break LOCA analysis of record assumes a single failure of one Emergency Diesel Generator (EDG) and credits flow from one charging pump. The peak cladding temperature for the analysis of record is 1929 F compared to the acceptance criteria of less than or equal to 2200 F. Half of the flow from a charging pump is assumed to go to the RCS and half is assumed to flow out the break. It was identified on 03-31-04 that a single failure of a DC power bus would be more limiting than the currently assumed single failure of an EDG. The DC bus failure will result in the EDG failure to start and also cause a charging loop isolation valve to the RCS to fail closed on loss of DC power. The other charging flow path to the RCS is on the assumed broken RCS leg so it is assumed to be spilled to the containment resulting in no charging flow to the RCS. Therefore, no charging flow would be delivered to the RCS as assumed in the current small break LOCA analysis of record. This condition has not been analyzed. Engineering judgment indicates that in the event of a small break accident the 10 CFR 50.46 acceptance criterion of 2200 F may be exceeded if the charging flow is not credited in the analysis of record. Therefore, this condition is reportable pursuant to 10 CFR 50.46 and 10 CFR 50.72(b)(3)(ii)(B). The current conservatism in the analysis of record and previous Westinghouse SBLOCA non-design basis scoping analyses using the NOTRUMP code provide assurance the plant is operating in a safe manner and is operable. The NOTRUMP analyses included cases with no changing flow. The peak clad temperature for these analyses were significantly below the acceptance criteria of 2200 F by more than 200 F. Thus, if an analysis were performed for Waterford 3 with the NOTRUMP code there is reasonable assurance due to the large margins that acceptable results would be achieved without any charging flow. This report is based on conservative judgment and available information. This report may be retracted based on the results of further evaluation. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 41040 | 14 September 2004 21:04:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Hurricane Watch in St. Charles Parish | The following information was provided by the licensee via facsimile: Waterford 3 entered Unusual Event Emergency Classification due to the National Weather Service declaring a hurricane warning for St. Charles Parish at 1600 CDT. The licensee has notified the NRC Resident Inspector.
Waterford 3 is no longer in an unusual event (as of 0502 CDT) . Hurricane watch/warnings for St. Charles Parish has been lifted at 0400. The licensee notified the NRC Resident Inspector. Notified NRC Region 4 IRC, FEMA (Kevin Biscoe), NRR (Catherine Haney), IRD (Peter Wilson), DHS (Frank Griggs). | ||
| ENS 41617 | 19 April 2005 22:51:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Found at Waterford Unit 3 | Waterford Unit 3 shutdown on 4/17/2005 at approximately 0047 hrs to begin refueling outage 13. During performance of the Alloy 600 Program pressurizer "bare metal" inspection, a small amount of boric acid was discovered in the annulus around Pressurizer heater sleeves C4 and D2. Entergy had already planned to weld repair/replace the Alloy 600 pressurizer sleeves and remaining Alloy 600 small bore instrument penetrations using Alloy 690 materials during this refueling outage. This condition constitutes a degradation of the RCS Pressure Boundary and is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(A) with a required followup written Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(ii)(A). The NRC Resident Inspector has been notified. | 05000382/LER-2005-001 | |
| ENS 41659 | 2 May 2005 12:55:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | During Steam Generator Tube Inspection - Greater than 1% (106 Tubes) of the Tubes Were Found Defective | On May 2, 2005, with Waterford 3 in Mode 6, it was determined that Steam Generator 1 In-Service Inspection had identified more than 1 % defective tubes. Per Technical Specification table 4.4.2-C. this percentage of defective tubes is characterized as "C-3" and should be reported to the NRC. Accordingly, this event is being reported pursuant to 10CFR50.72(b)(3)(ii) as an 8-Hr Non-Emergency Degraded Condition. In addition, per Technical Specification 4.4.4.5, the final results of the Steam Generator tube inspection, which fall into category C-3, shall be reported in a special report within 30 days and prior to resumption of plant operation. Defective tubes will be plugged. The NRC Resident Inspector was notified. | Steam Generator | |
| ENS 41687 | 10 May 2005 12:55:00 | 10 CFR 26.73, Applicability | Fitness for Duty - Alcohol | A licensed employee had a confirmed positive for alcohol during a for-cause test. The employee's unescorted access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41733 | 28 May 2005 02:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to Osha Due to Onsite Fatality | In accordance with 10CFR50.72(b)(2)(xi), this report is being made to notify the NRC, via ENS, under a four hour report based on a planned notification to OSHA. Based on 29CFR1904.39(b)(5) and (b)(2) OSHA requirements for notification in case of a fatality at work, the below information will be provided to OSHA. At 04:00 (CDT) on 5/28105, the site decided to report the following information, although the cause of death has not yet been determined. An employee exited the work area and was found sitting in a chair in the office area slumped over, unconscious, no pulse, no breathing. Ventilation and CPR was initiated. An AED was utilized, and one shock was advised and administered. EMT's arrived (at 21:50 CDT) and transported (the) employee (via ambulance) to the hospital (at 22:07 CDT) without incident, while still performing CPR. The employee was not contaminated. The licensee notified the NRC Resident Inspector. | ||
| ENS 41954 | 28 August 2005 03:04:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Hurricane Warning | At 2204 CDT, an Unusual Event was declared due to a Hurricane warning in St Charles Parish, LA. The licensee does not have any Limiting Conditions for Operation on safety related equipment. No electrical grid disturbances currently exist. The licensee will be notifying the NRC Resident Inspector. The licensee notified the State and Local government agencies.
Waterford 3 will commence a plant shutdown to Mode 4 on August 28, 2005 at 1100 CDT. The goal is to be approximately 325 deg F reactor coolant system temperature with both trains of Shutdown Cooling in service (expected at approximately 0100 CDT 08/29/05) . This is approximately 2 hours before hurricane force winds are expected on site. Notified R4DO (Jones), NRR EO (Kuo), IRD Mgr. (Wilson), DHS (York), and FEMA (Canupp).
As a result of Hurricane Katrina, all offsite power was lost at 0800 CDT. The emergency diesel generators successfully started and loaded. Notified R4DO (R. Bywater), NSIR (Leach, Wilson), NRR EO (Kuo, Jung), FEMA (Kuzia), DHS (Stransky).
The Hurricane warning for St Charles Parish, LA, has been discontinued by the National Weather Service. Initiating condition, D UE-5, is no longer applicable. Waterford 3 remains in an Unusual Event for initiating condition C UE-1, loss of all offsite power. An agreement to relax 60 minute notification updates to State and local authorities remains in effect. The licensee has notified State and local authorities as well as the NRC Resident Inspector. Offsite power is available to the licensee switchyard. However, the voltage is too high for plant specification and the licensee remains disconnected from offsite power and on the emergency diesel generators.
Waterford 3 remains in an Unusual Event for initiating condition C UE-1, loss of all offsite power. An agreement to relax 60 minute notification updates to State and local authorities remains in effect. The communications system for Public Automatic Branch Exchange (PABX) is working internally, but not externally. Other areas where communications are lost are Emergency Operations Facility (EOF), Operations hotline, Reactor Auxiliary Building (RAB), and the Emergency Notification System (ENS). Waterford has use of an industrial hotline to a circuit in the Parish of St Charles, however, this is very limited in its capability. The Civil Defense communication system is still inoperable.
NRC exited monitoring mode at 1800 CDT on 9/6/05. Notified R4 IRC (Waig), R4DO (Powers), IRD (Leach), DHS (Cartlidge), FEMA (Kuzia).
At 17:57 CDT the licensee exited from the Unusual Event. Offsite power was restored and communications were reestablished with (1) a functional operational hotline, (2) a dedicated open line to NRC Headquarters Operations Center, and (3) having the plant PABX circuits routed through Little Rock, AK area code. The plant is holding in Mode 5 until further notice, however, this may change when the decision to (or not to) repair a check valve in the Safety Injection system is made. Notified R4DO (Powers), IRD (McGinty), NRR EO (Hackett), DHS (Sullivan), and FEMA (Kuzia).
This report has been updated to reflect the correct time of the licensee exiting their unusual event classification. Unusual event was exited at 1740 CDT not 1757 CDT. Notified R4DO (Powers), IRD (McGinty), NRR EO (Jung). | Reactor Coolant System Emergency Diesel Generator Shutdown Cooling | |
| ENS 42138 | 12 November 2005 02:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After Loss of All Condenser Circulating Water Pumps | Waterford 3 manually tripped the reactor at 20:34 (CST) on 11/11/05 due to lowering main condenser vacuum caused by a loss of all circ water pumps. Subsequently an Emergency Feedwater Actuation Signal (EFAS) was received due to low steam generator levels. The plant is currently being maintained in mode 3 with both Steam Generators being fed from the Auxiliary Feedwater system with Steam Generator levels in the normal operational band for mode 3. The EFAS has been reset. The plant will be maintained in Mode 3 while a Post Trip Review is performed. This report is submitted as required by 10CFR50.72. All control rods fully inserted on the manual trip. The main steam isolation valves are shut and the heat sink is through the atmospheric dump valves. The electrical grid is stable and plant power is from the startup transformer. No primary or secondary relief valves or safety valves lifted. The site was able to restart the B & D circ water pumps. There is no significant steam generator tube leakage. The licensee notified the NRC Resident Inspector.
This notification is an update to an immediate notification that was called into the NRC Operations Center at 04:48 (eastern time) on November 12, 2005 (Event # 42138). The event was reported as an 8-hour event. This update to the original notification is to communicate that, based on further review, the event notification should have been made within 4 hours since it was reportable within 4 hours as well as the 8-hour reporting criteria and therefore the call should have been made within the 4 hour period. The condition is reportable under four hour reporting criteria 50.72(b)(2)(iv)(B) for a manual trip of the plant in anticipation of receiving a RPS Trip with the Reactor critical. The condition is reportable within eight hour reporting criteria 50.72(b)(3)(iv)(A) for the automatic actuation of EFAS upon low steam generator levels. This update does not report a change to the event description reported. As mentioned in the original report, Waterford 3 manually tripped the reactor at 20:34 on November 11, 2005 due to lowering main condenser vacuum caused by a loss of all circulating water pumps. Subsequently an Emergency Feedwater Actuation Signal (EFAS) was received due to low steam generator levels. The plant was then maintained in Mode 3 with both steam generators being fed from the Auxiliary Feedwater System with steam generator levels in the normal operational band for Mode 3. The EFAS had been reset at the time of the original notification. The plant was maintained in Mode 3 while a Post Trip Review was performed. Further details of the event will be reported under the 60 day LER reporting criteria. The licensee notified the NRC Resident Inspector. R4DO (Clark) notified. | Steam Generator Feedwater Main Steam Isolation Valve Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42636 | 12 June 2006 18:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Manual Actions Determined Untimely for Fire Safe Shutdown | As part of an analysis to determine feasibility of manual actions, it was identified that three actions for Fire Area RAB 7 could not be performed within the times prescribed in the Post Shutdown Analysis. The subject fire area is subdivided by part height fire walls. Waterford 3 has an approved Appendix R deviation for the part height fire wall configuration. Assuming a fire in Fire Area RAB 7 Operator entry into that same fire area is required. Results of the analysis indicate that the conditions in Fire Zone RAB 7B rapidly exceed the habitability threshold and do not moderate before ten minutes, the time at which one of the three manual actions is required, given a fire in Fire Zone RAB 7A. Because the space becomes uninhabitable, the manual action in Fire Zone RAB 7A is not feasible. A continuous fire watch has been established within Fire Area RAB 7 as a compensatory measure. This condition is reportable within 8 hours pursuant to 10CFR 50.72(b)(3)(v)(A) as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition. The licensee notified the NRC Resident Inspector. | ||
| ENS 42672 | 28 June 2006 19:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss and Subsequent Restoration of Electrical Power to the Emergency Operations Facility | On June 28, 2006 at approximately 1415 (CDT), power was lost to the Waterford 3 SES Emergency Operations Facility (EOF) when a tree fell onto the power line feeding the EOF. Power was automatically restored to the EOF from the EOF backup diesel generator. Several minutes later, power was lost to the EOF again when the EOF backup diesel tripped. (Normal) power was restored to the facility at approximately 1550, and the EOF is operable. The backup EOF facility was available during this event. The NRC Senior Resident has been notified. | ||
| ENS 42860 | 7 September 2006 12:50:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid System Actuation | On September 7, 2006 at about 0750 CDST, an invalid actuation of B train Engineered Safety Features Actuation System (ESFAS) occurred during maintenance on ESFAS relay K110 B. The following B train Engineered Safety Features Actuation System (ESFAS) relays de-energized: K101, K102, K103, K108 and K109. The event is reportable per 10CFR 50.73(a)(2)(iv)(A) as a 60 day notification and is being reported to the NRC by phone within 60 days. The invalid actuations did not occur when the systems had already been properly removed from service nor after the safety function had already been completed. The 32B buss de-energized which resulted in loss of pressurizer heaters Bank B and Containment Fan Coolers B and D actuated by shifting to slow speed. There was no actual Emergency Core Cooling System (ECCS) system flows actuated. All of the affected equipment functioned as designed when the relays deenergized The NRC Senior Resident was informed of the event. | Emergency Core Cooling System | |
| ENS 44454 | 1 September 2008 04:00:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to Hurricane Warning Onsite | Site declared a notification of unusual event under HU6/ EAL 7. The site is predicted to experience hurricane force winds from Gustav. As a conservative measure, the plant shutdown was completed at 2232 CDT. The licensee has notified the NRC Resident Inspector, state and local emergency management agencies. Notified DHS (Hill) and FEMA (Blankenship).
The site has exited the Notification of Unusual Event (NUE) at 0021 CDT because 1) grid stability has been restored, 2) HU6/EAL 7 is no longer applicable (Hurricane Gustav is no longer affecting the site), and 3) the NUE for leakage greater than 25 gpm (SU7/EAL 2) is no longer applicable (entered at 0726 on 9/1/2008 and terminated at 0729 on 9/1/2008). The reactor is in mode-4 (Hot Shutdown). Notified the R4DO (Jones), NRR (Blount), IRD (Gott), DHS (Dub), and FEMA (Casto). | ||
| ENS 45445 | 19 October 2009 14:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Stuck Open Moisture Separator Heater Relief Valve | At approximately 0915 CDT on 10/19/09 a Waterford 3 Moisture Separator Heater shell-slide relief valve (RS-203B) inadvertently opened, causing reactor power to increase from 100% to approximately 100.27% Rated Thermal Power (RTP). Operations reduced Main Turbine-Generator load by approximately 26 megawatts to restore reactor power to less than 100% RTP. At approximately 0942 CDT, Operations commenced a rapid plant shutdown because the relief valve would not re-close. At approximately 0944 CDT, Operations manually tripped the reactor due to a low condenser hot well level, just prior to reaching the Condensate Pumps Trip setpoint, to avoid a loss of Main Feedwater event. The Plant Protection System (PPS) responded as designed, resulting in an uncomplicated reactor trip. Emergency Feedwater Actuation Signal (EFAS) was received on low Steam Generator level as expected from reactor trip at or near full power. Steam Generator levels remained above the EFAS injection level setpoint so that actual injection of Emergency Feedwater did not occur. No other PPS actuation occurred. The plant is currently being maintained in Mode 3. Waterford 3 plans to commence refueling outage (RF16) at this time, approximately 1 - 2 days earlier than scheduled. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 45476 | 4 November 2009 00:03:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Employee Tested Positive for Alcohol. | An employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's unescorted access to the plant has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector. | ||
| ENS 45526 | 29 November 2009 05:44:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Containment Minimum Pathway Leak Rate Was Exceeded | 10CFR50 Appendix J Local Leak Rate Testing determined the total 'as-found' containment minimum pathway leak rate exceeded the maximum allowable containment leak rate per the Containment Leakage Rate Testing Program. This was primarily due to three penetrations that could not be pressurized to full test pressure. The maximum allowable leakage was assigned to both valves in each penetration since the valves can not be tested individually. This condition is reportable as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded per 10CFR50.72(b)(3)(ii)(A). The condition was discovered with the plant in Mode 5. Corrective actions have already been completed and all penetrations, as well as total containment leakage, is well within limits established by the Containment Leakage Rate Testing Program. The licensee has notified the NRC Resident Inspector. * * * RETRACTION FROM JIM POLLOCK TO PETE SNYDER AT 1711 EST ON 12/2/09 * * * On November 29, 2009, Waterford 3 reported that the total as-found containment minimum pathway leak rate exceeded the maximum allowable containment leak rate per the Containment Leakage Rate Testing Program. This was primarily due to three penetrations that could not be pressurized to full test pressure. The maximum allowable leakage was assigned to both valves in each penetration since the valves could not be tested individually. This condition was determined to be reportable as a condition of the nuclear power plant, including its principle safety barriers, being seriously degraded per 10CFR72(b)(3)(ii). Subsequently, Waterford 3 has evaluated the leak rates for all three penetrations and determined that as-found leakage rates assigned to the penetrations should not have been recorded at 630,000 sccm each, but should have been recorded at lower values. After one valve was re-worked in Penetration number 48, a Local Leak Rate Test (LLRT) was completed. With the new information from the second LLRT, a revised as-found leak rate of 97 sccm was assigned to the un-worked valve. For Penetration number 10, an Engineering evaluation of the test results was completed. The evaluation determined that the as-found leakage from this penetration was 191,000 sccm. This was based on calculating the maximum air flow capacity that could be obtained from the Leak Rate Monitor test equipment and using the first LLRT's penetration only reaching a test pressure of 43 psig instead of the 44 psig full test pressure. For Penetration number 11, troubleshooting activities determined that one of the valves had no detectable leakage; however, a leak rate of 9,100 sccm was assigned based on its previous as-left leakage value. With the lower leakage rates, the total as-found containment minimum pathway leak rate is calculated to be 230,682 sccm, which is within the limit of 630,000 SCCM. Since Waterford 3 had not exceeded the maximum allowable containment leak rate, EN #45526 is being retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Deese). | ||
| ENS 45670 | 1 February 2010 16:15:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Test for an Illegal Substance | A contract supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. | ||
| ENS 45672 | 2 February 2010 20:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Primary Eof Unavailable Due to Failure of Ventilation Damper Position Indication | On February 2, 2010, at approximately 1400 CST, Entergy personnel manually switched the Emergency Operations Facility (EOF) Ventilation System Panel to the isolation mode. This manual switching to the isolation mode will cause the isolation dampers to close and the movement of the dampers closing can normally be heard. An indication light on the ventilation panel will illuminate when the dampers close. For the manual isolation on February 2, 2010, these indications did not occur. Waterford 3 personnel have made a judgment that the isolation dampers may not be closed due the lack of the aforementioned indications. The EOF ventilation system is not able to perform its isolation function if the dampers are not closed. The ventilation system and associated dampers are designed to provide protection from radiological and toxic chemical release. This report is being made because the time to restore the ventilation system to functional status is indeterminate at this time, until the actual condition of the ventilation system can be established. The back up EOF facility is judged to be functional based on January 12, 2010 communications check and facility readiness verifications on January 28, 2010. The licensee informed the NRC Resident Inspector. | ||
| ENS 45939 | 20 May 2010 20:35:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Transportation Accident | This is a Non-Emergency 4 hour notification being made in accordance with 10 CFR 50.72(b)(2)(xi), an Offsite Notification. On May 20, 2010, at approximately 1535 CDT Waterford 3 was notified by the Louisiana (LA) State Police that a transportation accident occurred involving a low level radioactive shipping container that was being carried by a tractor trailer truck located at mile marker 22 East on interstate highway I-12 near Walker, LA in Livingston Parish. The surface contaminated object (SCO) metal shipment container was being used by Miller Transfer Houston to transport a used Reactor Coolant Pump (RCP) motor from Waterford 3 nuclear plant in Killona, LA to Westinghouse Waltz Mill in Madison, PA. The shipment weighing approximately 145,000 pounds departed Waterford 3 at approximately 1230 CDT with identification number of 10-3044 and pre-shipment radiological survey information of contamination levels of less than or equal to 15,000 dpm/100cm2 and dose rates of less than or equal to 0.4 mR/Hr. The tractor trailer is on its side, just off the interstate, on the grass. The shipping container boxed cover lid has opened, exposing the RCP motor, showing that it is still wrapped with contamination protective covering and still fastened to the container base and trailer. The initial, basic radiological assessment performed by the LA State Police Hazmat technician supervisor resulted in no indication of hazardous levels of radiation dose rates. Waterford 3 dispatched a Radiation Protection field team who arrived at approximately 1840 CDT and surveyed the container and the contents. The survey results indicate no radiological breach, with smear results of contaminations no higher than background and radiation dose rates of no higher than 0.5 mR/Hr. There was no report of other vehicles involved. The driver was contacted by cell phone, who reported that he has been transported to the hospital emergency room for a CAT scan. The failure mode has yet to be determined. The licensee has notified the NRC Resident Inspector. | ||
| ENS 46077 | 7 July 2010 19:00:00 | 10 CFR 26.719, FFD Reporting requirements | Administrative Review Revealed Fitness for Duty Program Issues | Pursuant to Fitness For Duty and Fatigue Rule reporting requirements of 10CFR26.719(b)(2)(ii), Entergy is making this notification associated with violation of Waterford 3 working hour policy. This is a conservative report to proactively document the identified working hour issues. The site assembled a team for the purpose of reviewing working hours within the Waterford 3 Security Department to determine if the department is in compliance with procedure EN-OM-123 (Fatigue Management Program) and 10 CFR 26 Subpart I (Managing Fatigue). While this comprehensive review is still in progress, two examples have been identified associated with supervisory violations of the Fatigue Rule 'Working Hours.' This report is intended to capture these examples and to encompass any other examples that may be identified during this review process. The two examples of procedural violations are: Security supervisor had exceeded the 26 hours maximum in any 48 hour period by working 16 hours on 6/1/2010 and 11.5 hours on 6/2/2010 with no waiver in effect (10CFR26(d)(1)(ii)). Security supervisor had exceeded a 9 day period without a 34 hour break during the period of 5/16/2010 to 5/24/2010 with no waiver in effect (10CFR26(d)(2)(ii)). Security has verified the current watch bill meets the procedural and regulatory requirements. Security has implemented interim measures for the supervisors to perform a documented validation of work hour management system (PQ&S) data. Condition Report CR-WF3-2010-4156 has been initiated and entered into the Waterford 3 corrective action program. The ongoing comprehensive review will be completed. Waterford 3 has communicated this issue with the Waterford 3 NRC Senior Resident (Inspector), the NRC Region IV Security Branch Chief, and the NRC NRR Project Manager. (During) the week of July 26, a follow up call is planned with NRC Region IV associated with the working hour issues. | ||
| ENS 46308 | 4 October 2010 20:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Oil Sheen Enters Intake Structure from Upstream | This is a Non-Emergency 4 hour notification being made in accordance with 10CFR 50.72(b)(2)(xi), an Offsite Notification. On October 4, 2010, at 1500 CDT, the Waterford 3 Operations supervision was notified by the Outside Operations watchstander that there appeared to be a diesel oil film from upstream of the Mississippi River collecting on the upstream side of the Waterford 3 non-safety related Circulating Water intake weir. The Chemistry Department environmental expert went to investigate with the Outside watch and verified that a diesel film was washed upon the north wall of the weir at the shore line forming about a 50 square foot sheen of diesel. Safety systems are not impacted by this condition. At 1609, the National Response Center was notified of the discovery of the sheen at our intake from an unknown source upstream. The incident report number was 956006. The Coast Guard was notified and the Waterford 3 environmental chemist received a call from the Coast Guard requesting information. The information was confirmed that a diesel film was washed upon the north wall of the intake structure weir at the shore line forming about a 50 square foot sheen of what appears to be red colored diesel. At 1734, Louisiana Hazardous Materials Hotline was notified (IR# 10-05779). The licensee notified the NRC Resident Inspector. | ||
| ENS 46824 | 6 May 2011 07:09:00 | Other Unspec Reqmnt | Discovery of an After-The-Fact Unusual Event | Operators were making preparations to fill the Containment Spray System riser to support outage activities. At 0204 CDT, when Containment Spray Riser Isolation Valve CS-125A was opened, pressurizer level began to lower. The licensee suspects leak by of a valve in the Shutdown Cooling System. At 0214 CDT, the leak was stopped. Pressurizer level was lowered by 2.6%. After reviewing the event, the licensee determined that the leak rate was greater than 25 gpm which would have resulted in a declaration of an Unusual Event under EAL CU1. Since the event had concluded, no declaration was made. The licensee notified the NRC Resident Inspector. | Shutdown Cooling Containment Spray | |
| ENS 47361 | 20 October 2011 13:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Inoperable Emergency Feedwater Pump Potentially Impacts Rhr Safety Function | At 08:00 CDT on 10/20/2011 Essential Chilled Water Loop B was declared inoperable while turbine driven Emergency Feedwater Pump AB was out of service for planned maintenance. Operability of Essential Chilled Water Loop B was restored at 08:50. During this time period, the application of cascading technical specifications rendered motor driven Emergency Feedwater Pump B inoperable. The remaining operable Emergency Feedwater Pump A is a design rated 50 percent pump; therefore, this event could have prevented fulfillment of the Residual Heat Removal safety function. Offsite power and Train A safety related equipment and systems were verified operable. Essential Chilled Water Loop B was declared inoperable because Essential Chiller B failed to automatically restart from a load recycle. The shift crew took action to align Essential Chiller AB to restore operability to Essential Chilled Water Loop B. The cause for the failure of Essential Chiller B is not yet known. Restoration of operability to Essential Chilled Water Loop B concurrently restored operability to Emergency Feedwater Pump B. No plant transient or safety system actuations occurred. Plant operation continues stable at 100 percent power. The licensee notified the NRC Resident Inspector. | Feedwater Residual Heat Removal | 05000382/LER-2011-005 |
| ENS 47652 | 8 February 2012 16:47:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report Involving a Non-Licensed Employee Supervisor | A non-licensed employee supervisor had a confirmed positive for alcohol during random testing. The employee's access to the plant has been terminated and his badge deactivated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| ENS 47897 | 3 May 2012 21:13:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Both Trains of the Controlled Ventilation Area System Inoperable | This is a non-emergency 8 hour notification being made in accordance with 10CFR50.72(b)(3)(ii) Degraded or Unanalyzed Condition and 10CFR50.72(b)(3)(v) Prevented Safety Function. Both trains of the Controlled Ventilation Area System (CVAS) were considered inoperable from 1613 to 1641 CDT on 5/3/2012. Waterford 3 was performing planned maintenance on CVAS Train A with the train declared inoperable. Operations supervision was informed by maintenance personnel that an electrical conduit near the work area and not related to the work in progress had been found with a conduit connection unthreaded and separated. The contained wires could be seen and appeared intact. Investigation by engineering personnel determined that the conduit was associated with CVAS Train B. Operations declared CVAS Train B inoperable and entered Technical Specification (TS) 3.0.3 at 1613 hours. Operations personnel were already restoring CVAS Train A following completion of maintenance. After performing the planned retest, Operations declared CVAS Train A operable and exited TS 3.0.3 at 1641 hours. No actions were taken to commence a power reduction or reactor shutdown. The CVAS is required following a design basis accident to draw all exhaust air from the CVAS areas of the Reactor Auxiliary Building through HEPA and charcoal filters before discharge to the atmosphere. The CVAS areas contain equipment such as the shutdown cooling heat exchangers, safety injection pumps and other areas with containment penetrations. The CVAS is credited in accident analysis for limiting the radiological release under a Loss of Coolant Accident condition. Following repair of the conduit connection, Operations declared CVAS Train B operable at 1956 hours. Waterford 3 remained stable at 100% power during the condition. Both trains of offsite and onsite electrical sources, ECCS, and Containment Cooling safety systems remained operable during the condition. The licensee notified the NRC Resident Inspector.
This is a non-emergency notification from Waterford 3 retracting a previous notification. On 5/3/2012 at 2107 EDT, Waterford 3 communicated Notification Message EN 47897, pursuant to 10 CFR 50.72(b)(2)(ii), Degraded or Unanalyzed Condition and 10 CFR 50.72(b)(3)(v), Prevented Safety Function, stating that both trains of the Controlled Ventilation Area System (CVAS) were considered inoperable from 1613 to 1641 CDT on 5/3/2012. Subsequent engineering evaluation has determined that the as-found condition would not cause the inoperability of CVAS Train B. With one train operable, the plant remained within analyzed conditions and capable of fulfilling safety functions. As such, no condition existed which required reporting. Based on this information, Waterford 3 is retracting notification message EN 47897. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Clark). | Shutdown Cooling | |
| ENS 47957 | 23 May 2012 22:00:00 | 10 CFR 26.719, FFD Reporting requirements | Contraband Found Inside the Protected Area | On May 23, 2012, at approximately 1700 hours, a beer bottle was discovered in a trash can inside the protected area. The bottle was discovered during trash collection by housekeeping personnel who reported the discovery to their supervision. The bottle was determined to contain remnants of moisture and had an odor of beer, constituting the potential presence of alcohol. This report is submitted pursuant to 10 CFR 26.719 (b)(1) based on the presence of alcohol in the protected area. The NRC Senior Resident Inspector has been notified. We do not know if it was consumed in the Protected Area. It was found in the Construction Support Building, not in a Vital Area. This has been entered into the Corrective Action Program and investigation has been in progress". | ||
| ENS 48045 | 22 June 2012 18:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | National Response Center Notified of Ion Exchange Resin Shipment Leakage | This report is submitted pursuant 10CFR50.72(b)(2)(xi) due to a notification that was made to the National Response Center. At 1345 CDT on June 22, 2012, Waterford 3 was informed by NRC Region IV that there was a government report concerning a Waterford 3 radioactive waste shipment. A follow-up review determined that the receiver of the shipment (Energy Solutions) had notified the National Response Center of a condition stating that a shipment received at their facility in Tennessee on June 21, 2012 was spilling water. The notification was identified as Incident Report number 1015375. The container held UN 2910, Limited Quantity, low specific activity, ion exchange resin and did not exceed any reportable quantities. Waterford 3 is sending shipping personnel to Energy Solutions to validate the state of the shipment. The NRC Resident Inspector has been notified. | ||
| ENS 48192 | 14 August 2012 16:26:00 | 10 CFR 26.719, FFD Reporting requirements | Confirmed Positive Fitness for Duty Test | A licensed employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's unescorted access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified. | ||
| ENS 48251 | 29 August 2012 20:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Evacuation Route Unavailable Due to Flooding from Hurricane Isaac | This is a non-emergency notification being made under 10CFR50.72(b)(3)(xiii) due to the potential loss of emergency preparedness capabilities. At approximately 1530 CDT, Entergy Operations, Inc. was informed of a road closure which could impact the ability to perform a potential evacuation required by the Waterford 3 Emergency Plan. Louisiana Governor's Office of Homeland Security and Emergency Preparedness (GOHSEP) reported that US Highway 51 to Interstate 10 was closed due to flooding from elevated water levels in Lake Pontchartrain resulting from Hurricane Isaac. Alternate evacuation routes remain available. The licensee has notified the NRC Resident Inspector. | ||
| ENS 48255 | 30 August 2012 11:20:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Operations Facility and Emergency Notification System Phone Out of Service | This is a non-emergency notification being made under 10CFR50.72(b)(3)(xiii) due to the potential loss of emergency preparedness capabilities due to the effects of hurricane Isaac. At approximately 0735, the Emergency Operations Facility (EOF) diesel generator was discovered to be tripped and not available, leaving the primary EOF without AC electrical power and, therefore, unavailable since normal AC power was also lost due to weather conditions. Efforts to repair the EOF diesel generator have been initiated and personnel have verified the Alternate EOF is functional. State, local and federal agencies have been informed of the plan to use the alternate EOF in case of an emergency preparedness activation. With essential personnel being sequestered at the Waterford 3 power block, the critical EOF functions can be performed from the on-site Technical Support Center (TSC), which is housed within the Control Room ventilated envelope. At approximately 0620 CDT, Entergy Operations, Inc. Emergency Preparedness staff discovered that the NRC Emergency Notification Systems (ENS) was not working. NRC Headquarters was notified at approximately 0715 CDT and was given an alternate phone number to contact Waterford 3. The ENS phone had previously been verified functional at approximately 0330 CDT. The NRC Health Physics Network (HPN) telephone line was verified to be functional at 0922 CDT. The licensee notified the NRC Resident Inspector. | ||
| ENS 48357 | 28 September 2012 19:10:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 - Defective Masoneilan Transducer Model 8005N | This message is notification to the NRC, pursuant to 10 CFR 21.21(d)(3)(i) requirements, that the Vice President Operations at Waterford 3 was notified on September 28, 2012 at 14:10 CDT of a condition which will be conservatively reported as a defect under the rule. A written report to the NRC will follow within 30 days. The basic component that is subject to reporting is the Masoneilan I/P (current to pneumatic) Transducer Model 8005N. These transducers are utilized in safety related applications at Waterford 3. This condition has been corrected in the plant. Waterford has identified that the subject transducer fails to calibrate at the high end of its span. No defective components are currently installed. Waterford 3 is operating normally at 100% power. This identified condition caused no loss of safety function and had no impact on public health and safety. The licensee notified the NRC Resident Inspector. | ||
| ENS 48687 | 21 January 2013 21:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Low Steam Generator Level | At 15:51 CST, Waterford 3 experienced an uncomplicated automatic reactor trip from 84.5% reactor power. The actuations of the Reactor Protection System (RPS) and the Emergency Feedwater Actuation System (EFAS) resulted from Steam Generator #1 Low Level, which is at a nominal 27.4% narrow range setpoint. Safety systems responded as expected. All three (3) Emergency Feedwater Pumps started and injected into Steam Generator #1. Auxiliary Feedwater pump has been started, feeding both Steam Generators (#1 and #2) at levels above the EFAS low level setpoint. All control rods inserted by the automatic RPS actuation. Electrical power is being supplied from normal off-site power and condenser vacuum is available for Steam Generator heat removal via the Steam Dump Bypass Control system. There are no safety systems out of service or inoperable, nor any safety system TS (Technical Specification) LCO (Limiting Condition for Operation) actions entered. The cause of the Steam Generator #1 Low Level condition, and associated Reactor Trip, is under investigation. This event occurred during the initial power escalation from refuel outage RF18, after attempting to place C Heater Drain Pump (HDP), the first of three, into service. After starting, C HDP tripped for a reason not yet verified. Subsequently, based on initial Control Room operator observations, the Steam Generator #1 Main Feedwater control valve position was observed to be at 10-20% open, but with an open position demand signal of 100%. Main Feedwater response to the reactor trip (Reactor Trip Override) was as expected. The NRC Resident Inspector has been informed. * * * UPDATE FROM WILLIAM HARDIN TO PETE SNYDER AT 1645 EST ON 3/7/13 * * * The original reactor power level stated in the report should be 91% in lieu of 84.5%. This information has been changed in the event heading. The licensee notified the NRC Resident Inspector. Notified R4DO (Werner). | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod Emergency Feedwater Actuation System | |
| ENS 49067 | 24 May 2013 17:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Intake Barrier Being Struck by Oil Tanker | At 0835 CDT, the Entergy Transmission Operations Center notified the Waterford 3 control room that a crude oil tanker had struck the dolphins at the cooling water intake structure on the Mississippi River. There were 4 out of 5 dolphins damaged, with 3 of these having substantial damage. The dolphins are hard structures anchored around the cooling water intake structure (which provide) protection from river traffic. Waterford 3 operations was unaffected by this event and thermal power remains at 100%. The intake structure, including the Circulating Water System, was unaffected by this event. Possible near-term effects of the event are a loss of protective barrier between river traffic and the intake structure due to the physical damage to the dolphins and hazards to navigation due to the loss of the dolphin lights. At 1230 CDT, the United States Coast Guard was notified of this event in accordance with Waterford 3 procedures. This notification is subsequent to the notification of the United States Coast Guard per 10CFR50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector. | Circulating Water System | |
| ENS 49252 | 5 August 2013 19:30:00 | 10 CFR 26.719, FFD Reporting requirements | an Off-Duty Licensed Operator Was Arrested for Driving Under the Influence | A off-duty licensed operator was arrested while driving a vehicle under the influence of alcohol. This was determined to violate the Waterford Fitness for Duty program. The individual has been administratively removed from performing licensed duties pending investigation. The licensee has notified the NRC Resident Inspector. | ||
| ENS 49326 | 5 September 2013 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Loss of Auxiliary Component Cooling Water Systems | This is a non-emergency notification from Waterford 3. Conditions were discovered which appear to require immediate NRC notification per 10CFR50.72(b)(3)(ii)(B) due to both trains of Auxiliary Component Cooling Water (CCW) being inoperable several times since 7/27/2012. A deficiency was identified with Auxiliary Component Cooling Water valve ACCW-126A, which is a part of the ultimate heat sink system, associated with the inability to adequately close this valve manually, locally, in order to preserve Wet Cooling Tower A inventory during an accident. System operability requirements came into question and resulted in Operations declaring the system inoperable on 9/4/2013 at 1509 CDT, and entering Technical Specification (TS) LCO 3.7.3 and associated cascading TS. As part of the review for NRC reporting requirements associated with the inoperable CCW Train A, historical information was discovered that indicated the redundant train of CCW, Train B, was declared inoperable several times, while ACCW-126A was presumably in the degraded or inoperable condition, since ACCW-126A had last been rebuilt on 7/27/2012. This condition requires immediate reporting to NRC under 10CFR50.72(b)(3)(ii)(B), an unanalyzed condition that significantly degrades plant safety. The condition was corrected and the system declared operable on 9/5/2013 at 1231 CDT, exiting the CCW TS LCO and associated cascading TS. Prior to the condition with ACCW-126A being corrected, a snubber pin (FWSR-60) was found missing on 9/5/2013 at 1228 CDT that could have adversely affected the ability to feed Steam Generator #2 with Emergency Feedwater. Less than 4 hours later, the pin was replaced, which restored the path to operable at 1609 CDT on 9/5/2013. The plant remained stable at 100% during this time. Plant risk index was 10.0 green. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 49875 | 5 March 2014 15:12:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Impact on Unfused Direct Current Ammeter Circuits in the Main Control Room | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined that a similar condition is applicable to the Waterford 3 Nuclear Station resulting in a potentially unanalyzed condition with respect to 10 CFR 50 Appendix R requirements. The original plant wiring design and associated analysis for an ammeter measuring current from the train AB Class 1E battery to its associated power distribution panel does not include overcurrent protection features to limit the fault current and is routed through multiple fire areas. The ammeter is located on the train AB power distribution panel in the train AB switchgear room. In the postulated event, a fire could cause one of the ammeter wires to short to ground. Simultaneously, it is postulated that the fire could cause another DC wire from the opposite polarity on the same battery to also short to ground. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (i.e., heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). There is no effect on plant operation. Fire watches have been implemented for affected areas of the plant as an interim compensatory measure. The licensee notified the NRC Resident Inspector.
Subsequent engineering evaluation has determined that the circuit for an ammeter measuring current from its Class 1E battery to its associated power distribution panel is not routed through multiple fire areas. Therefore, the IER 13-54 related condition is not, and was not, an unanalyzed condition at Waterford 3 that significantly degraded plant safety, and thus not required to be reported under 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector. Notified R4DO (Gaddy). | ||
| ENS 49925 | 17 March 2014 12:37:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Sirens | At approximately 0737 (CDT) on March 17, 2014, Waterford 3 was informed that four emergency sirens were inoperable. One other siren had previously been determined to be inoperable on March 12, 2014. Subsequent review during preparation of this notification has identified an additional eight inoperable sirens, which brings the total number of inoperable sirens to thirteen. There are a total of seventy-three sirens distributed among two parishes (counties). The loss of these thirteen sirens for more than one hour is considered a major loss of offsite response capability and is reported pursuant to 10 CFR 50.72(b)(3)(xiii). The two affected parish Emergency Operations Centers were notified of the condition and it was confirmed that they will use the preplanned alternative method of Route Alerting for the affected areas until notified that repairs to the sirens have been completed. Waterford 3 is working to repair the sirens. There is no effect on the plant. This issue has been entered into the Waterford 3 Corrective Action Program and appropriate corrective actions will be developed. The NRC Resident Inspectors, local agencies, and the State of Louisiana were notified. | ||
| ENS 49978 | 31 March 2014 20:58:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Six Emergency Sirens | At approximately 15:58 hours (CDT) on March 31, 2014, Waterford 3 was informed that five emergency sirens were inoperable during the performance of a monthly siren inspection. Subsequent review during preparation of this notification has identified an additional one inoperable siren, which brings the total number of inoperable sirens to six. There are a total of seventy-three sirens distributed among two parishes (counties). The loss of these six sirens for more than one hour is considered a major loss of offsite response capability and is reported pursuant to 10 CFR 50.72(b)(3)(xiii). The affected parish Emergency Operations Center was notified of the condition and it was confirmed that they will use the preplanned alternative method of Route Alerting for the affected areas until notified that repairs to the sirens have been completed. Waterford 3 has initiated preparations to repair the sirens, with actual repairs expected to commence tomorrow. The performance of the monthly siren inspection will continue during daylight hours today and resume tomorrow. There is no effect on the plant. This issue has been entered into the Waterford 3 Corrective Action Program and appropriate corrective actions will be developed. The NRC Resident Inspector and local agencies were notified. | ||
| ENS 50058 | 24 April 2014 15:22:00 | 10 CFR 26.719, FFD Reporting requirements | Non-Licensed Contract Supervisor Confirmed Positive on Follow-Up Fitness for Duty Test | A non-licensed contract supervisor was confirmed positive for alcohol on a follow-up fitness for duty test. The supervisor's access to the facility has been terminated. The licensee has notified the NRC Resident Inspector. | ||
| ENS 50191 | 18 April 2014 09:29:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Masoneilan Actuators Discovered to Fail-Closed Instead of the Required Fail-Open Configuration | This is a non-emergency notification from Waterford 3 required under 10 CFR PART 21 concerning the deviation of a basic component from manufacturing purchase specifications. On 04/18/2014, during pneumatic actuator replacement for Emergency Feedwater (EFW) valve EFW-224A (EFW to Steam Generator #1 Primary Flow Control Valve), it was identified that the fail action of the actuator was incorrect in that it was in a fail-closed instead of fail-open configuration. The actuator model, as identified in Waterford 3 site receipt/acceptance document and the actuator's name plate data, was a Masoneilan model 47 Sigma-F, which is specified as a fail-open actuator. However, the actuator was discovered to have a fail closed action, which is indicative of a Masoneilan model 48 actuator. The fail-action of these Masoneilan Sigma-F actuators can be changed in accordance with its technical manual TDM120.0565 with no additional parts; therefore, Waterford 3 corrected the fail action of EFW-224A's replacement actuator (to fail-open) and ensured the handwheel indicator was corrected prior to entering mode 4. Two of these actuators were received and accepted as Masoneilan model 47 Sigma-F (fail-open) actuators in September of 2010, both of which were labeled as such, yet configured in the fail-closed action. One of these two actuators remains in the warehouse and has been placed on hold. The vendor, Dresser Masoneilan, has been notified in writing on May 12, 2014. On 6/9/2014 at approximately 1600 CDT, Entergy concluded that for those potential applications of this valve operator model in the EFW system, had they been installed without correcting the deviation, it could have resulted in a substantial safety hazard in that it could have prevented fulfillment of EFW safety function, and therefore reportable under 10CFR PART 21. The Waterford 3 Site VP was informed on 06/11/2014. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 50378 | 18 August 2014 13:53:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Ultimate Heat Sink System Inoperable | Conditions were met that require immediate NRC notification per 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(D) due to both trains of the Ultimate Heat Sink (UHS) system being inoperable for approximately 83 minutes. This condition resulted from a planned system outage of train B Component Cooling Water (CCW), as a subsystem of the Train B UHS, and an unexpected trip of a Train A Auxiliary Component Cooling Water (ACCW) Wet Cooling Tower (WCT) fan, which is also a subsystem of the Train A UHS. CCW Train B was declared inoperable at 0820 CDT and removed from service for a planned relay replacement, resulting in entry of 72 hour shutdown Technical Specification (TS) LCO 3.7.3, and associated cascading TS on Train B, including TS LCO 3.7.4 Action a. Subsequently, at approximately 0853 (CDT), the 6A ACCW WCT fan tripped, rendering the redundant Train A UHS inoperable, causing entry into 1 hour TS LCO 3.7.4 Action b, which states, 'With both UHS trains inoperable, restore at least one UHS train to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.' At approximately 0948 CDT the WCT Fan 6A electric motor thermal overload relays were reset, the fan restarted, and operated properly. At 1016 CDT, CCW train B had been restored from the planned maintenance and was declared operable, exiting TS LCO 3.7.3 and associated cascading TSs on Train B. Although a plant shutdown was not commenced following expiration of the one hour specified in TS LCO 3.7.4 ACTION b, Operations was preparing for a power reduction and TS required plant shutdown. During this reported condition, an outside air intake valve of the Control Room Emergency Filtration system was inoperable for planned maintenance, for which TS 3.7.6.1 mitigating actions were in place. This valve was restored to Operable at 1210 CDT. Other safety systems remained available. The plant remained stable at 100% during this time. The NRC Resident Inspector was notified. | Control Room Emergency Filtration System | |
| ENS 51116 | 3 June 2015 22:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Main Feedwater Pump | This is a non-emergency notification from Waterford 3. At 1705 (CDT) the reactor was manually tripped in anticipation of an automatic trip due to loss of main feedwater pump 'A'. The plant is currently in mode 3 and stable with emergency feedwater feeding and maintaining both steam generators due to an automatic emergency feed actuation signal. During the trip, the 'B' electrical safety and non safety busses did not automatically transfer from the unit auxiliary transformer to the startup transformer causing a loss of off-site power to the 'B' electrical busses. This resulted in a loss of main feedwater pump 'B'. The 'B' emergency diesel generator started as designed and reenergized the 'B' safety related buses. The plant entered the emergency operating procedure for loss of main feedwater. Off-site power has been restored to the 'B' safety and non safety busses, and the emergency diesel generator 'B' is secured. All control rods fully inserted into the core following the trip. Decay heat is being removed by the main condenser using the turbine bypass valves. The electric plant is in a normal shutdown lineup. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Main Condenser Control Rod | |
| ENS 51295 | 6 August 2015 17:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Non-Functional Emergency Sirens | At 1200 CDT, on August 6th, 2015, during the monthly test of Emergency Plan sirens in Saint Charles Parish (county), all sirens failed to actuate when required. There are a total 73 sirens of which 37 sirens are in Saint Charles Parish, which covers approximately 49 percent of the total population within the Waterford 3 10-mile emergency planning zone. The siren vendor is currently investigating to repair the issue. Waterford 3 still maintains 100 percent backup notification capabilities. The Saint John the Baptist Parish Sirens remain fully operational. The Saint Charles Parish Emergency Manager has been notified. This event did not result in any adverse impact to the health and safety of the public. The Waterford 3 and Saint Charles Parish Emergency Response Plans include back-up processes to provide warning to affected areas, if required, in the event of the loss of sirens. The NRC Resident Inspector has been notified. | ||
| ENS 51348 | 26 August 2015 12:40:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Emergency Diesel Generators Declared Inoperable | This is a non-emergency notification from Waterford 3. On August 26, 2015, at 0111 CDT, Emergency Diesel Generator (EDG) 'A' was declared inoperable following a trip of EDG 'A' on Generator Differential. Technical Specification (TS) 3.8.1.1 actions b. and d. were entered. EDG 'A' was being routinely run in accordance with OP-903-115, 'Train A Integrated Emergency Diesel Generator/Engineering Safety Features Test', Section 7.4, '24 hr EDG A Run with Subsequent Diesel Start' to satisfy Technical Specification Surveillance Requirement 4.8.1.1.2 6. EDG 'B' was subsequently started per TS 3.8.1.1 action b. (1). At 0740 CDT, EDG 'B' was declared inoperable and TS 3.8.1.1 f. was entered due to the exhaust fan not starting when the diesel engine was started. Troubleshooting determined that the EDG B exhaust fan did not start due to HVR-501B (EG B ROOM OUTSIDE AIR INTAKE DAMPER) not opening. Action was taken to isolate air and fail HVR-501B to its open safety position. At 1001 CDT, EDG 'B' was declared operable and TS 3.8.1.1.f. was exited following verification of proper operation of the EDG 'B' exhaust fan. Waterford 3 is currently in TS 3.8.1.1 actions b. and d. Actions to verify a temporary EDG is available and restore EDG 'A' to operable status are in progress. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v) (A) and 10 CFR 50.72 (b)(3)(v) (D), 'event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (A) shut down the reactor and maintain it in a safe shutdown condition' and (D) 'mitigate the consequences of an accident due to both emergency diesel generators being inoperable.' The NRC Resident Inspector has been notified.
The following is a correction to a non-emergency event notification from Waterford 3 originally made on 8/26/2015: On August 26, 2015, at 0111 CDT, Emergency Diesel Generator (EDG) 'A' was declared inoperable following a trip of EDG 'A' on Generator Differential. Technical Specification (TS) 3.8.1.1 actions b and d were entered. EDG 'A' was being routinely run in accordance with OP-903-115, 'Train A Integrated Emergency Diesel Generator/Engineering Safety Features Test,' Section 7.4, '24 hr EDG A Run with Subsequent Diesel Start' to satisfy Technical Specification Surveillance Requirement 4.8.1.1.2(e)6. EDG 'B' was subsequently started per TS 3.8.1.1 action b.(1). At 0740 CDT, EDG 'B' was declared inoperable and TS 3.8.1.1 f was entered due to the room exhaust fan not starting when the diesel engine was started. Troubleshooting determined that the EDG B room exhaust fan did not start due to HVR-501B (EDG B ROOM OUTSIDE AIR INTAKE DAMPER) not opening. Action was taken to isolate air and fail HVR-501B to its open safety position. At 1001 CDT, EDG 'B' was declared operable and TS 3.8.1.1.f was exited following verification of proper operation of the EDG 'B' room exhaust fan. Waterford 3 is currently in TS 3.8.1.1 actions b and d. Actions to verify a temporary EDG is available and restore EDG 'A' to operable status are in progress. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(A) and 10 CFR 50.72(b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function of structures or systems that are needed to (A) shut down the reactor and maintain it in a safe shutdown condition and (D) mitigate the consequences of an accident due to both emergency diesel generators being inoperable. The NRC Resident Inspector has been notified. Notified R4DO (Warnick) | Emergency Diesel Generator | |
| ENS 51447 | 4 October 2015 04:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 2307 CDT Waterford 3 experienced an automatic reactor trip and all Control Element Assemblies (CEAs) inserted into the core. The cause of the automatic reactor trip is currently under investigation. The plant is currently in Mode 3 (Hot Standby) and stable with Main Feedwater feeding and maintaining both Steam Generators. Main Feedwater Pump 'A' tripped subsequent to the reactor trip. Emergency Feedwater actuated following the plant trip as expected, but was not required to maintain Steam Generator level. The plant entered the Emergency Operating Procedure for an uncomplicated reactor trip and has now transitioned to the normal operating shutdown procedure. Unit 3 is in a normal post trip electrical lineup. The Main Condenser is in-service removing decay heat.. The licensee informed the NRC Resident Inspector. | Steam Generator Feedwater Main Condenser | |
| ENS 51464 | 9 October 2015 23:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Inadvertent Siren Activation During Testing | At 1800 (CDT) on October 9, 2015 a polling test was initiated in Saint Johns Parish to test the circuitry of the installed sirens. During the polling test no sirens are expected to sound as it is only a circuitry test. Siren number SJ39 inadvertently sounded for 15 to 20 minutes, and no others. Saint Johns Parish notified the parish residents that the sounding of the siren was inadvertent via a Parish wide cable television channel and a press release. A contract vendor has disabled the siren and will troubleshoot and repair starting on October 12, 2015. All remaining sirens within Saint Johns Parish remain operational and capable of being activated when required. 0% of the population is affected by the loss of this siren due to siren overlap. Time to repair and restore siren SJ39 to service is still being investigated. This event is reportable pursuant to 10CFR 50.72 (b)(2)(xi), News Release or Notification of Other Government Agency. The NRC Resident Inspector has been notified. |